Evaluation of SCALE, Serpent, and MCNP for Molten Salt Reactor applications using the MSRE Benchmark

The International Reactor Physics Benchmark Experiment's MSRE benchmark provides zero-power critical validation data that has been used to assess the accuracy and consistency of SCALE, MCNP, and Serpent for design and licensing of a modern MSR. The codes were used to model the benchmark and a wide range of variations in the geometry, nuclear data library, code versions, and problem specifications. Most of these cases demonstrated excellent consistency, within two standard deviations of the stochastic error, for the reactivity and associated reactivity worth of the variation. However, the codes over-predicted the initial criticality of the MSRE by 3%, which is larger than the provided experimental uncertainty. It was shown that care must be taken to ensure that the bound scattering treatment is consistent with comparing codes. It was also shown that for the MSRE, the use of the ENDF/B-VII.0 library introduced a significant increase in the reactivity (> 200 pcm). Code predictions for two coefficients of reactivity (fuel and isothermal) were validated with data from the MSRE and were well within the experimental uncertainty, thus providing confidence in each code to provide accurate data for safety analysis calculations. These results provide quantifiable estimates of the computational variation one could expect due to the choice of any one of these codes, or a particular nuclear data library, for both initial criticality and reactivity coefficients.